Text

Safety Insp Repts 50-456/87-25 & 50-457/87-24 on 870713-0806.Violations Noted.Major Areas Inspected:Startup Test Results Evaluation
	ML20238A703
Person / Time
Site:	Braidwood
Issue date:	08/25/1987
From:	Vandenburgh C, Wright G; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20238A682	List: IR 05000456/1987025; ML20238A679; ML20238A690;
References
	50-456-87-25, 50-457-87-24, NUDOCS 8708310178
	Download: ML20238A703 (13)
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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Reports No.' 50-456/87025(DRS); 50-457/87024(DRS)

Docket Nos. 50-456;60-457 Licenses No. NPF-37; CPPR-133 Licensee:

Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name:

Braidwood Station, Units 1 and 2 Inspection at:

Braidwood Site, Braidwood, Illinois Inspection Conducted: Jul 13 through August 6, 1987 ff$ $]

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Inspector-C.

. VanD bu gh Date

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l Approved By:/ G. Wright, Chief Test Program Section Date

Inspection Summary Inspection on July 13 through August 6. 1987 (Reports No. 50-456/87025(DRS);

No. 50-457/87024(DRS))

Areas Inspected:

Routine, announced safety inspection of startup test results evaluation (Modules 72596 and 72600).

Results:

In the area inspected, three violations were identified: Failure to iaentify deficient conditions (Paragraph 2.a), failure to evaluate test results (Paragraph 2.e.(2)), and failure to perform testing to incorporate l

the requirements of design documents (Paragraph 2.h).

Because the violation

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concerning the failure to identify deficient conditions discussed in l

Paragraph 2.a meets the requirements of 10 CFR 2, Appendix C, Part V, Section A, l

a violation will not be issued.

Each of the violations was determined to have minimal safety significance; however, the failure to perform testing'in accordance with design documents, to adequately evaluate the test results, and to document and resolve apparent deficient conditions encountered during testing is of concern.

The results of the inspection determined that the licensee had adequately performed these functions on testing perf-ormed subseouent to the identified discrepancies, and as such the violations are' considered to oe isolated' cases.

In addition, one unresolved item concerning the long term corrective actions required for identified seal' table leakage was identified (Paragraph 2.e.1).

8708310178 870825 l

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DETAILS 1.

Persons Contacted Commonwealth Edison Company (CECO)

'*E. Fitzpatrick, Station Manager

D. Shamblin, Project Manager

' *K. Kofron, Production Superintendent C. Schroeder, Services Superintendent

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L. Davis, Assistant Superintendent

D. Paquette, Assistant Superintendent Maircenance

J. Bauer, Test Review Board Coordinator

P. Barnes, Regulatory Assurance Supervisoi

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T. Simpkin, Regulatory Assurance l

M. Takaki, Regulatory Assurance J. Doyle, Regulatory Assurance

R. Aker, Radiation / Chemistry Supervisor

G. Orlov, Production Manager Staff Assistant

D. Kapinus, Assistant Technical Staff Supervisor

R. Mertogul, Assistant Technical Staff Supervisor

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G. Vanderheyden, Acting Training Supervisor

R. Kyrouac, Quality Assurance Superintendent

J. Gosnell, Quality Control Supervisor

R. Legner, Senior Operating Engineer

L. Kepley, Operating Staff

M. Miller, Operating Staff J

M. Gorski, Nuclear Safety Engineer s

S.'Zepinski, System Test Engineer

T. Coomer, System Test Engineer j

T. Weiss, Project Engineering Department j

D. Elias, Project Engineering Manager j

J. Valentine, Project Engineering Department P., Oresheack, Project Engineering Department Westir,ghouse

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J. Linnard, Senior Nuclear Engineer l

B. Palowitch, Startup Test Supervisor

U.S Nuclear Regulatory Commission

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C. VanDenburgh, Senior Reactor Inspector

T. Tongue, Senior Resident Inspector

Denotes those personnel present at the exit interview of August 6,1987.

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2.

Startup Test Results Evaluation

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The inspector reviewed the results of the below listed startup test

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procedures to verify'that all test changes were identified and approved in accordance with administrative procedures; all test deficiencies were-i appropriately resolved, reviewed by management and retested as required;

test results were evaluated by appropriate engineering personnel and

l specifically compared with acceptance criteria; data was properly I

recorded, signed, dated, and documented as test deficiencies if out i

of tolerance; and test results were approved by appropriate personnel:

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Core Load Sequence j

BwSU FH-30, " Initial Core Load Sequence" l

Post Core Loading Precritical Test Sequence BwSU EM-30A, " Pipe Vibration (Pre Critical)"

BwSU IT-32A, " Thermal Power Measurement and Statepoint i

Data Collection (Pre-Critical)"

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BwSU RC-31A, " Reactor Coolant Flow Measurement" I

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BwSU RC-32,

" Reactor Coolant Flow Coastdown"

BwSU RC-33, " Reactor Coolant System Leak Test" I

BwSU RD-31,

" Rod Control System Checkout"

BwSU RD-33,

" Rod Drop Time Measurement Testing" BwSU RP-30, " Reactor Protection Logic"

BwSU RY-30,

" Pressurizer Sprays, Heaters and Bypass l

Flow Adjustments"

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Initial Criticality and Low Power Test Sequence

BwSU FW-30A, " Steam Generator Controller Response at 3-6% Power"

BwSU IC-31, "Incore Flux Mapping at Low Power" Collection (0% Power)"

BwSU LM-30,

" Reactor Lotse Parts Monitor (0% Power)"

BwSU NR-31,

" Initial Criticality"

BwSU NR-34C, " Operational Alignment of Excore Nuclear Instrumentation (Frior to Initial Criticality)

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BwSU NR-34D, " Operational Alignment of Excore Nuclear Instrumentation i

(FH-32, Initial Criticality and Low Power Test Sequence)

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30% Power Test Sequence

BwSU IT-32C, " Thermal Power Measurement and Statepoint Data

Denotes those procedures for which final Project Engineering Department (PED) approval has not been received.

This item will be followed as an open item pending inspector receipt end evaluation (456/87025-01).

a.

With respect to the results evaluation of RwSU FH-30, the inspector noted the following concerns:

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e (1) A note entered at 0252 hours0.00292 days 0.07 hours 4.166667e-4 weeks 9.5886e-5 months on October 27, 1986, in the Sequence of Events' Log (SEL), Appendix A, indicates that Fuel Assembly A555 in core location H-8 appears to be bowed.

Further discussions with the licensee indicate that this condition was evaluated and accepted by the licensee's vendor (Westinghouse)

following discovery.

A subsequent memo from Westinghouse (RCH-129-86) dated October 29, 1986, documented the acceptability of the bowing in Assembly H-8 and the adjacent assembly J-8.

Although the bowing of Assembly H-8 and the effect on luading adjacent assemblies warranted further evaluation and acceptance by the vendor, a'Startup Test Deficiency was not initiated to provide a resolution to this discrepancy.

Additionally, the inspector noted that the Sequence of Events Log, the Test Evaluetion and the Test Review Board Comments did not address the resolution to this problem.

(2) An entry at 2250 hours0.026 days 0.625 hours 0.00372 weeks 8.56125e-4 months on October 30, 1986, in the SEL indicates

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that core loading was halted after the discovery that a grease

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cap on the coupling between the Speed Control Generator and Refueling Machine Gear Box was missing.

The SEL indicates that a search for the missing grease cap was discontinued at 0420 hours0.00486 days 0.117 hours 6.944444e-4 weeks 1.5981e-4 months and that core loading was recommenced at 1938 hours0.0224 days 0.538 hours 0.0032 weeks 7.37409e-4 months i

on October 31, 1986.

Further discussions with the licensee l

indicate that a Deviation Report, DVR 20-186-028, was initiated

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to document and provide corrective actions for this occurrence.

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As corrective action, the licensee halted fuel loading and performed visual inspections of the refueling cavity in an effort to locate the missing grease plug.

Westinghouse provided a revised safety evaluation of the effects of a potential foreign object in the Reactor Coolant System and concluded that the missing grease plug should not affect any

accident previously analyzed or create any new accidents.

Although the missing grease cap warranted the interruption of fuel loading and significant corrective actions in the

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form of a DVR and an evaluation by Westinghouse, a Startup Test

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Deficiency was not initiated to identify the corrective actions taken in response to this event.

Additionally, the inspector noted that the SEL, the Test Evaluation and the Test Review Board Comments did not discuss the corrective actions taken for this deficiency.

(3) Special Test Condition 7.3.5.2 required that the Containment Evacuation Alarm be verified operational and logged at eight hour intervals in Appendix C, Table C1.

The SEL indicated that core alterations were in progress between 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on October 26, 1986, and 0252 hours0.00292 days 0.07 hours 4.166667e-4 weeks 9.5886e-5 months on October 27, 1986; however, the SEL and Tabla C1 do not indicate that the Containment Evacuation Alarm was verified operational between 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months and 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on October 26, 1986.

The inspector noted that the SEL, the Test Evaluation, and the Test Review Board Comments do not resolve this apparent discrepancy.

Further discussions l

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with the licensee indicated that in the course of their review of the test results of BwSU FH-30, the Project Engineering Department (PED), has also identified that initial Condition 7.3.5.2 was not satisfied; a Startup Test Deficiency was not initiated as required; and that the Test Evaluation and Test Review Board Comments did not identify this deficiency.

In response to this concern, the Braiduood Station determined l

that the Containment Evacuation Alarm was actually verified l

operational at 1719 hours0.0199 days 0.478 hours 0.00284 weeks 6.540795e-4 months on October 26, 1986, and logged in

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the unit operator's log, but not the Appendix C, Table C1 log as required by the test procedure.

In each of the above occurrences, a Startup Test Deficiency was l

not initiated to document a significant event or discrepancy which occurred during the performance of the test procedure.

As a result of the failure to initiate test deficiencies, the results of subsequent evaluations and corrective actions were not identified within the test results package. Although adequate corrective actions were separately undertaken in the first two events, the test results package does not document the resolution to the deficient condition.

In the third example, corrective actions were not undertaken because the deficient condition was not identified.

The Braidwood Startup Test Manual defines deficiencies and requires that the STE complete the Startup Deficiency Form BwSM 4-5 to document the deficient condition, the recommended resolution, and the corrective action taken.

The licensee has indicated that several testing inadequacies, including the failure to properly initiate deficiencies, were

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identified following the performance of BwSU FH-31, " Post Core Loading Precritical Test Sequence."

In Technical Staff Letter TS-87-044, dated May 7, 1987, a list of recommendations for improving the overall

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quality of startup testing was provided to all testing personnel.

I This letter identified forty-one separate recommendations to improve l

the test program quality.

Item 4.3 discussed the need to initiate a deficiency whenever a test step cannot be completed as written or if an unexpected result occurs. Training was provided for all testing personnel on May 13 and 14, 1987, to address the good testing practices and " lessons learned" from the BwSU FH-31 sequence.

The inspector also noted, based on a review of test procedures performed since this date, that test deficiencies are being properly I

identified in the test procedure and in the subsequent Test Review

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Board (TRB) Comments. There are numerous examples where test deficiencies were initiated at the request of the TRB and where

deficiency resolutions were required to be clarified to provide a

detailed explanation of the corrective actions undertaken.

The failure to initiate Startup Test Deficiencies and thereby

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document the corrective actions for the occurrences identified above

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is a violation of 10 CFR 50, Appendix B, Criterion XVI, " Corrective i

Action," as implemented by the Commonwealth Edison Quality Assurance

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Manual and the Braidwood Startup Manual.

However, because the violation discussed meetr ihe requirements of 10 CFR 2, Appendix C, Part V, Section A, in thW.

the violation was identified by the l

licensee; it fits a Severity Level IV or V; it was not required I

to be reported; it was adequately corrected; and it was not a violation which was expected to be prevented by the corrective action to a previous violation, a violation will not be issued.

The inspector has no further concerns concerning this test or this occurrence, and will continue to monitor the startup test results l

evaluations to ensure that good testing practices and " lessons

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learned" identified in the Technical Staff Letter TS-87-044 are incorporated.

b.

With respect to the result evaluation of BwSU EM-30A, the inspector noted that in the Project Engineering Department (PED) approval of the test results a recommendation was made to clarify the requirements of Test Step 9.0.5 as they relate to +he performance of.a worst case vibration measurement for future testing.

BwSU EM-30 will be performed at various power levels throughout the Startup Test Sequence and as such this recommendation will be followed as an open item to ensure the test requirements are modified (456/87025-02).

c.

With respect to the results evaluation of BwSU RC-31A, the inspector noted that although the Reactor Coolant System loop flows meet the-minimum acceptance criterion of greater than 90% of the thermal design flow, the loop flows measured by differential pressure transmitters 1FT-414, 1FT-415, IFT-436, 1FT-444, and 1FT-446 exceeded the expected value for the Reactor Coolant Pump mechanical design flow.

This condition has been reviewed by Westinghouse in l

Letter CBW-5877 dated April 9, 1987, and preliminarily approved.

This item will be followed as an open item pending the review and final approval of the full power calorimetric flow measurement test

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BwSU RC-31B in the 100% Power Test Sequence (456/87025-03).

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d.

With respect to the results evaluation of BwSU RC-32, the inspector noted that the Low Flow Trip Time Delay was exceeded because the measured time for Control Rod Drive Mechanism Stationary Gripper Coil release time (Tg) exceeded the the expected value of 150 milliseconds.

The test procedure calculated Tg as the time from when the fastest reactor trip breaker opened until the Stationary Gripper Coil voltage decayed to zero.

After consultation, the licensee redefined the i

measurement of Tg as the time from when the fastest reactor trip l

breaker opened until the Digital Rod Position Indicator (DRPI)

i voltage demonstrated that the control rod had begun to drop into the l

core.

Westinghouse confirmed that the second method is a practical measurement of the Stationary Gripper Coil release time in tetter CBW-5920 dated May 26, 1987.

The inspector has verified that this methodology is consistent with the requirements of the recommended Westinghouse Startup Test Procedures and the Final Safety Analysis Report (FSAR) testing requirements.

Using the revised methodology for measurement of Tg, an acceptable time for

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the Low Flow Trip Time Delay was demonstrated.

The inspector has

no further concerns with this test.

j e.

With respect to the results evaluation of BwSU RC-33, the inspector noted the following coments:

(1) Test Step 9.1.14 required a visual inspection of the Reactor Coolar,t System and documentation in Data Sheet 11-1 to satisfy the requirement of Accepte.nce Criteria 4.1.1 for no pressure boundary leakage.

Data Sheet 11.1, Sub-System:RC-1 (Reactor Pressure Vessel), and Startup Test Deficiency 1-C-1 document that leakage of approximately one drop every five to ten minutes (one-half teaspoon in five and one-half hours) was observed at the Incore Seal Table (Connections H6 and 83).

Deficiency 1-C-1 has been resolved as acceptable because the leakage observed is considered to be Identified Leakage vice Pressure Boundary Leakage as specified in the Technical Specifications definitions.

Therefore the acceptance criterion requiring no pressure boundary q

leakage has been satisfied.

i Further review identified that Nuclear Work Request, NWR A07144, had been initiated to inspect, reassemble and tighten the high pressure fittings of the Incore Seal Table after it was

discovered that several high pressure seals were loose following i

core loading.

In the process of inspecting and reassembling all of the high pressure seals, 50 of the 58 seals were determined to be loose.

Following this maintenance, the high pressure fittings of the Incore Seal Table were inspected at i

normal operating pressure during the performance of BwSU RC-33 on March 8, 1987.

This inspection identified boron crystal i

formation at Fittings H4, H6, N6, A9, and C5 and water

accumulation at B3 and H6.

Another inspection was performed on May 27, 1987, which identified moist threads and accumulation of crystals at Fittings B3, H6, and J10.

As a result of the determination of the acceptability of the leakage identified at

Fittings B3 and H6 in Startup Test Deficiency 1-C-1, NWR A07144 was closed.

Although the leakage amounts identified at the Incore Seal Table are considered acceptable in terms of the Technical Specification leakage criterion, the inspector remains concerned that the original deficient condition for which NWR A07144 was initiated (i.e., loose fittings at the Incore Seal Table) has not been adequately resolved.

Further action should be undertaken to ensure that the identified leakage does not increase and that when plant conditions warrant, further actions are taken to correct the leakage identified.

This item will be followed as an unresolved item pending further licensee evaluation of the corrective actions necessary for this deficiency (456/87025-04).

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(2) Test Step 9.1;11 required a leak test of the Reactor Coolant System by the performance of Technical Specification

i Surveillance IBWOS 4.6.2.1.d-1, "RCS Water Inventory Balance."

Startup Test Deficiency 3-D-1 documented that the pressurizer s

temperature recorded during the test exceeded the temperature /

density values provided in Table 8, " Saturated fluids / Gases,"

I and that the extrapolated value for pressurizer fluid density used in the calculation of the RCS Water Inventory was incorrect.

The deficiency resolution provided a new value for fluid density l

and recalculated the test results to demonstrate that the

acceptance criteria for Identified and Unidentified Leakage was satisfied.

The inspector was not able to verify the fluid densit;, provided by the deficiency resolution.

Further review by the licensee determined that the extrapolated value for water density provided in the resolution of Deficiency 3-D-1 was also incorrect.

Deficiency 3-D-1 provided a density which was derived by assuming a 23% steam quality in the pressurizer.

This assumption was

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based.upon an extrapolation of the data contained in Table B

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and the Standard Steam Tables.

The licensee determined that the correct values for pressurizer fluid densities are calculated l

based on the difference of the steam and water densities because steam replaces the water as level decreases.

The assumption of a 23% steam quality does not provide the correct density required

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for the calculations of IBWOS 4.6.2.1.d-1.

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The correct value for water density corresponding to the i

pressurizer temperature which was measured in the test is j

conservative (i.e., smaller) than the value provided in the s

resolution of Deficiency 3-D-1.

Therefore, there is no' safety l

significance with respect to the use of the incorrect value.

i However, the failure to accurately calculate the RCS Water

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Inventory as a result of an incorrect deficiency resolution is a violation of the 10 CFR 50, Appendix B, Criterion XI, " Test Control," requirements to evaluate test results to ensure that

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test requirements have been satisfied (456/87025-05).

f.

With respect to the results evaluation of BwSU RD-31, the inspector noted that Acceptance Criteria 4.8 was not satisfied in that D.C. Hold Cabinets 2AC, IBD, and 2BD did not supply the expected LATCH and HOLD currents to the Stationary Gripper Coils of the selected rod control groups.

This discrepancy was identified and resolved by Startup Test Deficiency 3-A-1, which required a Project Engineering Department j

(PED) evaluation of the measured currents in Action Item Request,

AIR R20-87-054.

Westinghouse provided a partial review of the test j

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data in Letter CBW-5868 dated April 2, 1987.

This item will be j

followed as an open item pending the final Project Engineering

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Department review and approval of the test results (456/67025-06).

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With respect to the results evaluation of BwSU RD-33, the inspector I

noted that rod drop times were determined for all rods.

Three rods j

(B04, B12, and J03) were determined to have rod drop times that

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exceeded the two-sigma acceptance criteria limit determined from the average rod drop times.

The rod drop times for these rods were measured an additional three times as required.

The inspector noted that acceptance criteria for the additional rod drop times were not provided.

Following the results evaluation of. Rod Drop Testing performed at Byron Generating Station Unit 2, Westinghouse recommended

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that acceptance criteria be provided for additional rod drops and for

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the proper operation of the decelerating devices (i.e., dashpot region)

for future testing performed at the Braidwood Generating Station.

Because these recommendations were not incorporated into BwSU RD-33 performed at Braidwood Unit 1, the Project Engineering Department approval of the BwSU RD-33 test results will be followed as an open item to ensure that the additional rod drop times are acceptable and that the decelerating devices performed satisfactorily (456/87025-07).

As a result of the inspection activities performed at Byron Generating Station, the inspector has reviewed several PED and Westinghouse results reviews which have made recommendations for incorporation into future testing at Braidwood.

In order to ensure that these lessons learned are incorporated into the Braidwood Units 1 and 2 Startup Test Programs, the licensee should ensure that all Byron Generating Station Unit 2 test results approvals are reviewed.

This concern will be followed as an open item pending further review to t

ensure that an effective lessons learned program has been implemented (456/87025-08).

h.

With respect to the results evaluation of BwSU RP-30, the inspector noted that Startup Test Deficiency 1-C-1 was initiated in Test Steps 9.9.22 and 9.9.25 because the logic function for Position C=23 (Undervoltage Diode Output Module) of Train 2 of the Reactor Protection System should have been 1/3 rather than the 1/4 specified in the test procedure.

Test Change Request TCR-1, which approved a major change to the test procedure, altered the logic of the Undervoltage Diodes and Output Module from 1/3 to 1/4 due to a Test Procedure Open Item (TP01).

Adequate justification for the TCR and the TP01 were not provided on either document.

The TPOI was initiated in response to a concern identified in Inspection Report No. 50/456-86052.

In Paragraph 8.a(5), the inspector noted that the logic for the Undervoltage Diodes and Output Module was not consistent between Trains 1 and 2 of the Reactor Protection System.

As a result of the TPOI, TCR-1 was incorrectly incorporated to resolve the apparent discrepancy with the logic testing.

Based on discussions with the licensee's staff, the inspector determined that the appropriate vendor's reference material (Westinghouse Solid State Protection System Vendor's Manual) was not consulted to verify the correct logic prior to the incorporation of TCR-1.

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Following the completion of BwSU RP-30 with the incorrect logic, the licensee's staff identified that the logic tested was in error and that the test procedure verification of the 1/4 logic was apparently performed incorrectly.

Subsequently, the licensee corrected the logic in TCR-4 and satisfactorily reperformed the logic testing.

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The failure to provide adequate justification for a major change to the test procedure; the subsequent approval of an incorrect major change to the test procedure; and the inadequate verification of the test procedure step which resulted in the verification of an incorrect reactor protection logic are considered to be violations of the

10 CFR 50, Appendix B, Criterion XI, " Test Control," requirements to perform testing in accordance with written test procedures which incorporate the requirements of applicable design documents

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Although the licensee's review identified and corrected the incorrect logic testing and as a result there was no direct safety significance to this specific occurrence, the failure to adequately review and approve a major test change to the test procedure as well as the failure to accurately verify the test procedure as written are significant violations of NRC requirements which could have safety

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significance if not corrected.

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With respect'to the results evaluation of BwSU RY-30, the inspector l

noted that Test Section 9.3 demonstrated that the Pressurizer Heater's effectiveness was guater than anticipated in the first forty seconds of the increasing pressure transient.

This deviation from Acceptance Criteria 4.3 was documented in Startup Test Deficiency 3-F-1 and resolved by the preliminary approval of Westinghouse in I

Letter CBW-5845 dated March 27, 1987, and the Project Engineering j

Department in a letter dated April 3, 1987.

The inspector has no I

further concerns with respect to this test.

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With respect to the results evaluation of BwSU LM-30, the inspector noted that retesting of accelerometer Channels 2, 11 and 20 is required following troubleshooting and repair identified in Startup Test Deficiencies 1-B and 1-D.

The required repairs will be performed by Nuclear Work Requests NWR 13698, A14036 and A14069.

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following these repairs is tracked by Action Item Requests AIR R20-87-115 and R20-87-122.

The required retesting of these components will be followed as an open item (457/87025-10).

k.

With respect to the results evaluation of BwSU NR-31, the inspector noted that the test procedure did not verify a minimum neutron count level of two counts per second and a minimum signal to noise ratio of two as required by Regulatory Guide 1.68.

The licensee demonstrated that the verification of minimum Source Range neutron l

count level was performed by Test Steps 9.1.1 and 9.7.1 by establishing a minimum counting interval such that the lowest reading Source Range channel is greater that 1000 counts in less than 500 seconds.

This step effectively verifies that a minimum count rate of two counts per second is achieved.

The licensee also demonstrated that the verification of minimum signal to noise ratio was performed in Test Step 7.3.3.14 of BwSU FH-32, " Initial l

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1 Criticality and Low Power Test Sequence." The inspector has no further concerns with this test, i

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With respect to the results evaluation of BwSU NR-34D, the inspector j

noted that Acceptance Criteria 4.1 required that at least one and

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one-half decades of overlap be verified between the Source and Intermediate Range Nuclear Instrumentation channels.

The test procedure required that overlap be recorded during the Initial Criticality Startup until the Source Range Nuclear Instruments j

were manually blocked.

The inspector noted that the data recorded i

in Data Sheets 11.7 and plotted in Figures 2, 3, 4, and 5 of

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Appendix C document overlaps of less than one decade because the Source Range Instruments were blocked before the required overlap was achieved.

The licensee reviewed the test results and determined that although the test results could be extrapolated to demonstrate the required overlap, sufficient data had not been recorded to demonstrate the overlap.

Accordingly, Action Item j

Request AIR R20-87-126, was initiated to require Section 9.7 of i

BwSU NR-340 be reperformed in BwSU TG-30, " Test Sequence at 30%

Power." The licensee has performed Section 9.7 and provided the results for review.

This data clearly demonstrates greater than one and one-half decades of overlap.

In order to prevent the occurrence of a similar problem on Unit 2, a Test Procedure Open Item was initiated to require the Source Range Instrument to remain unblocked until sufficient overlap data has been recorded.

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The inspector has no further concerns with this test.

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With respect to the results evaluation of BwSU IT-32C, the inspector noted that Startup Test Deficiency 401 A-E identified that the Wide

Range Reactor Coolant System temperatures recorded on Data i

Sheets 11.1.1G were outside the expected ranges and that RTD ITR-433A l

and the WR ID WR T COLD Conditioning Card exceeded their allowable tolerances.

The corrective action identified for these deficiencies indicated that Nuclear Work Request NWR 14467 was initiated to replace RTO 1TR-433A.

The deficiency also indicates that retesting is not required following these replacements.

This item will be followed as an open item pending the Project Engineering Department review l

of the test results and deficiency resolutions (456/87025-11).

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Open Items l

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Open items are matters which have ' Deen.aiscussed with the licensee, which

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will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both.

Open items disclosed during the inspection are discussed in Paragraphs 2, 2.b, 2.c, 2,f, 2.g, 2.j, and 2.m.

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Unresolved Items

Unresolved items are matters about which more information is required

in order to ascertain whether they are acceptable items, violations,

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or deviations.

Unresolved items disclosed during the inspection are discussed in Paragraphs 2.e.1).

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5.

Exit Interview I

The inspector met with' licensee representatives denoted in Paragraph 1 at the conclusion of the inspection on August 6, 1987.

The inspector summarized the purpose and scope of the inspection and the findings.

The inspectors also discussed the likely informational content of.the inspection report with regard to documents or processes reviewed by

,

the inspectors during the inspection.

The licensee did not identify any such documents / processes as proprietary.

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IR 05000456/1987025

Text

REGION III

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